Published August 1971 | Version v1
Journal article

Effect of Zircaloy-clad fuel rod failure on BWR core spray cooling

  • 1. General Electric Co., San Jose, Calif. (USA)

Description

The first full-length, electrically heated, 49-rod, Zircaloy-clad simulated BWR fuel bundle with internally pressurized rods was spray cooled under loss-of-coolant conditions. The heater rods were internally pressurized with argon to simulate fission product gas inventory. Many perforations and severe rod distortions occurred near the center of the bundle. Nevertheless, spray cooling, initiated at a maximum cladding temperature of 1920°F, was effective in limiting cladding temperatures to ~2250°F. Electrical failure of 10 heater rods complicated interpretation of the results, and it is estimated that the maximum temperature would have been ~2360°F had the failures not occurred. The maximum coolant flow area reduction around a single rod caused by local perforations was 50%. However, this flow area reduction did not appreciably impair the effectiveness of the spray cooling system. That is, analysis performed using current General Electric (GE) loss-of-coolant technique s and heat transfer coefficients derived from stainless-steel-clad bundle tests predicted the maximum cladding temperature to within 20°F.

Additional details

Identifiers

Publishing Information

Journal Title
Nuclear Technology
Journal Volume
11
Journal Issue
4
Series
Nucl. Technol.
Journal Page Range
532-543
ISSN
0029-5450

INIS

Country of Publication
United States
Country of Input or Organization
United States
INIS RN
3016004
Subject category
S22: GENERAL STUDIES OF NUCLEAR REACTORS;
Resource subtype / Literary indicator
No Abstract
Descriptors DEI
BOILING WATER REACTORS; DEFORMATION; ECCS; FAILURES; FUEL CANS; FUEL RODS; HEAT TRANSFER; LOSSES; MOCKUP; PERFORMANCE; REACTOR ACCIDENTS; SPRAY; WATER COOLANT; ZIRCALOY
Descriptors DEC
ACCIDENTS; FUEL ELEMENTS; REACTOR MATERIALS; REACTOR SAFETY; RODS; SPUTTERING; WATER COOLED REACTORS

Optional Information

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